The present invention relates generally to methods and apparatus which monitor nuclear reactors, and more particularly to such methods and apparatus employing parameters which are monitored exterior of the reactor core.
As is well known, conventional pressurized water reactors typically contain a reactive region, commonly referred to as the "core", in which sustained fission reactions occur to generate heat. The core includes a plurality of elongated fuel rods comprising fissile material, positioned in assemblies and arranged in a prescribed geometry governed by the physics of a nuclear reaction. Neutrons bombarding the fissile material promote the fissionable reaction which, in turn, releases additional neutrons to maintain a sustained process. The heat generated in the core is carried away by a cooling medium which circulates among the fuel assemblies, and is conveyed to heat exchangers which, in turn, produce steam for the generation of electricity.
A neutron absorbing element is also included within the cooling medium in controlled variable concentrations to modify the reactivity, and thus the heat generated within the core as required. In addition, control rods are interspersed among the fuel assemblies, longitudinally movable axially within the core, to control the reactivity of the core and thus its power output.
The power output, or distribution within the core is determined by, among other things, neutron flux distribution. While the radial power distribution of the core is fairly predictable, due to the prescribed arrangement of fuel assemblies and the positioning of control rods which are symmetrically situated radially throughout the core, the axial power distribution can vary greatly during reactor operation. Furthermore, over the lifetime of a nuclear power plant, changing fuel management schemes can result in significant changes in both the magnitude and distribution of neutron flux and, hence, neutron fluence throughout the reactor vessel beltline region. It is, therefore, desirable to monitor the axial power distribution within the core of a nuclear power plant.
Axial power distribution monitoring is typically conducted within conventional nuclear power plants through in-core instrumentation as well as ex-core instrumentation. In-core instrumentation systems are generally comprised of movable miniature fission chambers which are designed to yield information pertaining to neutron flux distribution at selected locations in the reactor core, fuel assembly outlet thermocouples, and in some cases, fixed miniature fission chambers. Such systems provide an accurate measure of the core relative power distribution, but provide no automatic protective function for the reactor.
On the other hand, ex-core instrumentation systems are typically comprised of uncompensated, long ion chambers or power range detectors which are located in four vertical instrument wells outside of the reactor vessel and symmetrically placed with respect to the core. Such power range detectors are calibrated to their respective in-core system and are used to provide automatic reactor protection against adverse power peaking. Additional information is also obtained by the use of supplementary passive neutron dosimeters installed within the annular-shaped reactor cavity between the walls of the reactor vessel and the primary biological shield.
Prior art approaches which have utilized such supplementary passive neutron dosimeters have typically hung the dosimeters by stainless steel, nickel or iron wires at various locations within the reactor cavity. Accurate placement of the dosimeters, however, was difficult at best since the reactor cavity in most nuclear power plants is narrow and often largely inaccessible. Movement of the dosimeters, due to various causes such as mechanical vibrations, heavy ventilating air currents passing over the dosimeters, and expansion or contraction of the reactor vessel during heat-up and cool-down, further complicate the accuracy and repeatability of dosimeter placement over the life of the nuclear power plant. It is, therefore, readily apparent that a method and apparatus for accurately and repetitively placing supplementary passive neutron dosimeters within the reactor cavity would be desirable.
Another problem with such prior art approaches to the placement of supplementary passive neutron dosimeters is the interference they often create with refueling operations. During a typical refueling operation (i.e., during the replacement of the fuel assemblies of a nuclear reactor after exhaustion of their fuel), the head assembly of the reactor vessel must be removed in order to withdraw the spent fuel assemblies. Additional shielding must be provided, however, due to the potentially dangerous radiation levels which are experienced during refueling. Accordingly, the reactor cavity is sealed off in order that the space above the reactor vessel may be flooded with water.
Such sealing of the reactor vessel is accomplished most often in one of two ways. A first approach merely clamps a heavy steel plate over the reactor cavity, with the steel plate including gaskets on either side of the cavity. The second approach utilizes the steel plate of the first approach, but additionally employs an inflatable bladder which serves to further seal the reactor cavity at a top portion thereof. As can be readily appreciated, therefore, any reactor cavity dosimetry system which is to be used in such nuclear power plants must avoid the possibility of puncturing the inflated bladder.
In addition to the access problems presented by the varying configurations of reactor cavities, problematic placement of a reactor cavity dosimetry system is further aggravated by the need for personnel changing the dosimeters to wear protective clothing. As is conventional, in order to work about a deactivated reactor, personnel must wear many layers of protective clothing and a full-face respirator. Not only does such protective clothing severely impair the dexterity of the wearer, but it can also lead to the wearer's suffering heat prostration under long periods of use. It would, therefore, be desirable in the design of a reactor cavity dosimetry system to provide one which is capable of rapid and remote deployment.
A recent method of deploying a reactor cavity dosimetry system is disclosed in Ser. No. 032,894, to which the present application is a continuation-in-part. The system and method disclosed therein, however, is dependent upon the remote positioning and retrieval of the supplementary passive neutron dosimeters from the reactor sump which is located beneath the reactor vessel. High radiation levels or deep water within the sump, as well as the lack of permanently installed scaffolding therein, all lead one to the conclusion that, when such conditions are present, a reactor cavity dosimetry system would be more desirably positioned and retrieved from atop the reactor vessel.